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At least 1,225 records · Page 68

Natural Transmutation of Actinides via the Fission Reaction in the Closed Thorium–Uranium–Plutonium Fuel Cycle

It is shown for a closed thorium–uranium–plutonium fuel cycle that, upon processing of one metric ton of irradiated fuel after each four-year campaign, the radioactive wastes contain ~54 kg of fission products, ~0.8 kg of thorium, ~0.10 kg of uranium isotopes, ~0.005 kg of plutonium isotopes, ~0.002 kg of neptunium, and “trace” amounts of americium and curium isotopes. This qualitatively simplifies the handling of high-level wastes in nuclear power engineering.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Delayed Gamma-ray Spectroscopy for Safeguards Applications

The delayed gamma-ray assay technique utilizes an external neutron source (D-D, D-T, or electron accelerator-driven), and high-resolution gamma-ray spectrometers to perform characterization of SNM materials behind shielding and in complex configurations such as a nuclear fuel assembly. High-energy delayed gamma-rays (2.5 MeV and above) observed following the active interrogation, provide a signature for identification of specific fissionable isotopes in a mixed sample, and determine their relative content. Potential safeguards applications of this method are: 1) characterization of fresh and spent nuclear fuel assemblies in wet or dry storage; 2) analysis of uranium enrichment in shielded or non-characterized containers or in the presence of a strong radioactive background and plutonium contamination; 3) characterization of bulk and waste and product streams at SNM processing plants. Extended applications can include warhead confirmation and warhead dismantlement confirmation in the arms control area, as well as SNM diagnostics for the emergency response needs. In FY16 and prior years, the project has demonstrated the delayed gamma-ray measurement technique as a robust SNM assay concept. A series of empirical and modeling studies were conducted to characterize its response sensitivity, develop analysis methodologies, and analyze applications. Extensive experimental tests involving weapons-grade Pu, HEU and depleted uranium samples were completed at the Idaho Accelerator Center and LLNL Dome facilities for various interrogation time regimes and effects of the neutron source parameters. A dedicated delayed gamma-ray response modeling technique was developed and its elements were benchmarked in representative experimental studies, including highresolution gamma-ray measurements of spent fuel at the CLAB facility in Sweden. The objective of the R&D effort in FY17 is to experimentally demonstrate the feasibility of the delayed gamma-ray interrogation of shielded SNM samples with portable neutron sources suitable for field applications.

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Nuclear Material Accounting and Control (NMAC) for a Commercial Fuel Debundling Facility in the United States

The concept of fuel debundling proposes that all non-fuel components of a fuel assembly be removed prior to beginning the recycling process. It was introduced to reduce the quantity of and cost of disposing high level radioactive waste during recycling in the United States. However, nuclear material accountancy and control programs could also benefit from fuel debundling due to a reduction in self-shielding material for individual fuel rods compared to a fuel assembly. Current NRC regulations and guidance point to an independent fuel debundling facility with a material-possession-only license as the best option for implementing this concept into a used fuel recycling scheme in the United States. This allows the facility to comply only with simple NMAC requirements under Subpart B of 10 CFR Part 74 without the need to submit an NMAC plan to the NRC, exempting it from more tedious and complex NMAC requirements. However, a recycling facility may desire a Pu mass estimate from measurements of each individual fuel rod to improve accountancy into the dissolver. Many technologies and methodologies were considered for performing NMAC in an independent fuel debundling facility in the United States. Few were able to meet anticipated throughput demands of up to 3,300 rods/day and achieve measurement uncertainties suitable for input accountancy for a recycling facility. The best option was a combination of passive gamma-ray spectroscopy and neutron coincidence counting to meet all requirements for the proposed Fuel Debundling Detector (FUDD). Follow-on work will include a full uncertainty analysis to inform what assumptions can be made to meet input accountancy objectives before beginning experimental activities.

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Irradiation pretreatment for coal desulfurization

Process using highly-penetrating nuclear radiation (Beta and Gamma radiation) from nuclear power plant radioactive waste to irradiate coal prior to conventional desulfurization procedures increases total extraction of sulfur.

Hsu, G. C.↗

Manufactured soils for plant growth at a lunar base

Advantages and disadvantages of synthetic soils are discussed. It is pointed out that synthetic soils may provide the proper physical and chemical properties necessary to maximize plant growth, such as a toxic-free composition and cation exchange capacities. The importance of nutrient retention, aeration, moisture retention, and mechanical support as qualities for synthetic soils are stressed. Zeoponics, or the cultivation of plants in zeolite substrates that both contain essential plant-growth cations on their exchange sites and have minor amounts of mineral phases and/or anion-exchange resins that supply essential plant growth ions, is discussed. It is suggested that synthetic zeolites at lunar bases could provide adsorption media for separation of various gases, act as catalysts and as molecular sieves, and serve as cation exchangers in sewage-effluent treatment, radioactive-waste disposal, and pollution control. A flow chart of a potential zeoponics system illustrates this process.

Ming, Douglas W.↗

Exploring the speciation of actinide salts in the presence of contaminants related to molten salt reactors

Molten salt reactors (MSR) are up-and-coming Generation IV nuclear reactors with either coolant and/or radioactive fuel in molten salt form. Due to their improved safety, efficiency, affordability and convenient waste processing system, these reactors are considered a superior alternative to conventional ones. However, prior to commercial implementation, it is necessary to have an extensive understanding of the reactions occurring in the MSR. The safety parameters of the MSR are determined by the rheological properties such as density and viscosity of the molten salt and these properties are governed by the local structure of the molten salts. Due to the high operational temperatures of MSR's, corrosion plays an important role. Because of their influence, the container corrosion products become part of the fuel salt matrix. In this work, we are investigating how these corrosion products, e.g., Mo, Ni, Cr, would impact the speciation of the actinide metal centers. For the initial studies, lanthanides are used as a surrogate for actinides. Lanthanide chloride salts (LnCl3) are mixed with transition metals and excess of alkali or alkali earth metal salts or salt mixtures (LiCl, NaCl, KCl). These are added to an alumina crucible, placed in a quartz tube, and sealed under a vacuum. This is heated up to around 1000 °C and slowly cooled to a temperature above the melting point of the salt and at that temperature, the reaction is taken out of the furnace. The excess salt is removed, and the resulting product is analyzed using X-ray diffraction techniques, scanning electron microscopy, and spectroscopic methods such as Raman, UV-Visible, and FTIR spectroscopy.

37 - INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL C↗

Privatization of solid waste collection services: Lessons from Gaborone

Highlights: • We compared efficiency and effectiveness of waste collection by the public and private sector. • Public sector performs better than private sector in some areas and vice versa. • Outsourcing waste collection in developing countries is hindered by limited capacity on contractual issues. • Outsourcing collection in developing countries is hampered by inadequate waste information. • There is need to build capacity in the public sector of developing countries to support outsourcing. - Abstract: Formal privatization of solid waste collection activities has often been flagged as a suitable intervention for some of the challenges of solid waste management experienced by developing countries. Proponents of outsourcing collection to the private sector argue that in contrast to the public sector, it is more effective and efficient in delivering services. This essay is a comparative case study of efficiency and effectiveness attributes between the public and the formal private sector, in relation to the collection of commercial waste in Gaborone. The paper is based on analysis of secondary data and key informant interviews. It was found that while, the private sector performed comparatively well in most of the chosen indicators of efficiency and effectiveness, the public sector also had areas where it had a competitive advantage. For instance, the private sector used the collection crew more efficiently, while the public sector was found to have a more reliable workforce. The study recommends that, while formal private sector participation in waste collection has some positive effects in terms of quality of service rendered, in most developing countries, it has to be enhanced by building sufficient capacity within the public sector on information about services contracted out and evaluation of performance criteria within the contracting process.

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Recent Regulatory Challenges Associated with Managing NORM and TENORM Waste Streams in the United States - 19609

Recent regulatory changes in several states in the United States (U.S.) have introduced new challenges to industries impacted with naturally occurring radioactive materials (NORM) and/or technologically enhanced NORM (TENORM) requiring innovative solutions. Because NORM and TENORM are specifically exempt from U.S. Nuclear Regulatory Commission (NRC) regulations, the onus falls on the 50 states to regulate their use and disposal. Over the past 10 years, regulations applying to NORM and TENORM have been drafted, passed, or updated in multiple states, including North Dakota, Pennsylvania, Illinois, and West Virginia to name a few. Solutions and lessons learned before and during the recent regulatory changes for a variety of clients, including specialty metals, mining, and oil and gas (O and G), are presented. In addition, the design and implementation of the comprehensive study of NORM/TENORM associated with conventional and unconventional O and G production activities in Pennsylvania on behalf of the Pennsylvania Department of Environmental Protection (PADEP) is presented, and the development of analytical solutions and waste processes designed to minimize the impact of the new regulations while maintaining compliance is discussed. (author)

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Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208 (Final Report)

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

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Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208 (Final Report)

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

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In-Situ Chemical Precipitation of Radioactive Liquid Waste - 12492

This paper presented in-situ chemical precipitation for radioactive liquid waste by using chemical agents. Results are reported on large-scale implementation on the removal of {sup 137}Cs, {sup 134}Cs and {sup 60}Co from liquid radioactive waste generating from Nuclear Research and Training Centre. Total amount of liquid radioactive waste was 35 m{sup 3} and main radionuclides were Cs-137, Cs- 134 and Co-60. Initial radioactivity concentration of the liquid waste was 2264, 17 and 9 Bq/liter for Cs-137, Cs-134 and Co-60 respectively. Potassium ferro cyanide was selected as chemical agent at high pH levels 8-10 according to laboratory tests. After the process, radioactive sludge precipitated at the bottom of the tank and decontaminated clean liquid was evaluated depending on discharge limits. By this precipitation method decontamination factors were determined as 60, 9 and 17 for Cs-137, Cs-134 and Co-60 respectively. At the bottom of the tank radioactive sludge amount was 0.98 m{sup 3}. It was transferred by sludge pumps to cementation unit for solidification. By in situ chemical processing 97% of volume reduction was achieved. Using the optimal concentration of 0.75 M potassium ferro cyanide about 98% of the {sup 137}Cs can be removed at pH 8. The Potassium ferro cyanide precipitation method could be used successfully in large scale applications with nickel and ferrum agents for removal of Cs-137, Cs-134 and Co- 60. Although DF values of laboratory test were much higher than in-situ implementation, liquid radioactive waste was decontaminated successfully by using potassium ferro cyanide. Majority of liquid waste were discharged as clean liquid. %97.2 volumetric amount of liquid waste was cleaned and discharged at the original site. Reduced amount of sludge transportation in drums is more economical and safer method than liquid transportation. Although DF values could be different for each of applications related to main specifications of original liquid waste, this study shows that in-situ treatment of liquid waste by using potassium ferro cyanide is not only a cost effective method but also reduce radiological risks as well. (authors)

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Analysis of Defense Waste Processing Facility Sample: Off-Gas Condensate Tank Sample Batch 27862

The Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (now Savannah River Mission Completion (SRMC)), through a Technical Task Request (X-TTR-H-00107), to characterize the “as-received” Off-Gas Condensate Tank (OGCT) Sample identified as sample batch 27862, which was delivered to SRNL Shielded Cells on April 30, 2021. This OGCT report is the third of three sample characterization reports dealing with this Defense Waste Processing Facility (DWPF) Recycle Diversion Project. The other two recycling stream characterization reports are the Recycle Collection Tank (RCT) and the Slurry Mix Evaporator Condensate Tank (SMECT) sample report.

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Why SRS Matters - H Canyon

A video series presenting an overview of the Savannah River Site's (SRS) mission and operations. Each episode features a specific area/operation and how it contributes to help make the world safer. This episode features H Canyon's mission and operations. H Canyon and its adjoining H B Line facility represent the last full-scale radio chemical processing capabilities left in the United States.

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Corrosion impact of reductant on DWPF and downstream facilities

Glycolic acid is being evaluated as an alternate reductant in the preparation of high level waste for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). During processing, the glycolic acid is not completely consumed and small quantities of the glycolate anion are carried forward to other high level waste (HLW) facilities. The impact of the glycolate anion on the corrosion of the materials of construction throughout the waste processing system has not been previously evaluated. A literature review had revealed that corrosion data in glycolate-bearing solution applicable to SRS systems were not available. Therefore, testing was recommended to evaluate the materials of construction of vessels, piping and components within DWPF and downstream facilities. The testing, conducted in non-radioactive simulants, consisted of both accelerated tests (electrochemical and hot-wall) with coupons in laboratory vessels and prototypical tests with coupons immersed in scale-up and mock-up test systems. Eight waste or process streams were identified in which the glycolate anion might impact the performance of the materials of construction. These streams were 70% glycolic acid (DWPF feed vessels and piping), SRAT/SME supernate (Chemical Processing Cell (CPC) vessels and piping), DWPF acidic recycle (DWPF condenser and recycle tanks and piping), basic concentrated recycle (HLW tanks, evaporators, and transfer lines), salt processing (ARP, MCU, and Saltstone tanks and piping), boric acid (MCU separators), and dilute waste (HLW evaporator condensate tanks and transfer line and ETF components). For each stream, high temperature limits and worst-case glycolate concentrations were identified for performing the recommended tests. Test solution chemistries were generally based on analytical results of actual waste samples taken from the various process facilities or of prototypical simulants produced in the laboratory. The materials of construction for most vessels, components and piping were not impacted with the presence of glycolic acid or the impact is not expected to affect the service life. However, the presence of the glycolate anion was found to affect corrosion susceptibility of some materials of construction in the DWPF and downstream facilities, especially at elevated temperatures. The following table summarizes the results of the electrochemical and hot wall testing and indicates expected performance in service with the glycolate anion present.

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Challenges in the UK Generic Design Assessment Process - 17062

In the UK, Generic Design Assessment (GDA) is a voluntary step-wise joint assessment process undertaken by the Office for Nuclear Regulation (ONR) and the Environment Agency (EA) of a generic reactor design. The process focuses purely on the reactor design and does not consider any site specific issues. The key objective of GDA is for the regulators to consider the adequacy of claims, arguments and evidence (CAE) put forward by a requesting party (RP) to demonstrate the safety of its design in the UK context. This includes all CAE relating to management of spent fuel and radioactive waste. The CAE are generally applied to a reference design put forward by the RP that is substantially complete in terms of the overall concepts and functions of the major systems, having already been deployed elsewhere. GDA assessment encompasses a broad range of topic areas which include fuel and core design, management of spent fuel, radioactive waste management and decommissioning. Other technical disciplines also have valid interests in particular aspects of the management of spent fuel. This paper summarises the UK GDA process and the types of issues raised; in particular aspects of the proposed operation of liquid waste management systems and Spent Fuel Interim Storage are presented and the instructive resolution of the RP's original limited design scope to allow the regulators to carry out a meaningful assessment. Some examples are presented and useful lessons learned for future GDAs are discussed. (authors)

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ELAWD GROUT HOPPER MOCK-UP TESTING

A 10-inch READCO mixer is used for mixing the premix (45 (wt%) fly ash, 45 wt% slag, and 10 wt% portland cement) with salt solution in the Saltstone Production Facility (SPF). The Saltstone grout free falls into the grout hopper which feeds the suction line leading to the Watson SPX 100 duplex hose pump. The Watson SPX 100 pumps the grout through approximately 1500 feet of piping prior to being discharged into the Saltstone Disposal Facility (SDF) vaults. The existing grout hopper has been identified by the Saltstone Enhanced Low Activity Waste Disposal (ELAWD) project for re-design. The current nominal working volume of this hopper is 12 gallons and does not permit handling an inadvertent addition of excess dry feeds. Saltstone Engineering has proposed a new hopper tank that will have a nominal working volume of 300 gallons and is agitated with a mechanical agitator. The larger volume hopper is designed to handle variability in the output of the READCO mixer and process upsets without entering set back during processing. The objectives of this task involve scaling the proposed hopper design and testing the scaled hopper for the following processing issues: (1) The effect of agitation on radar measurement. Formation of a vortex may affect the ability to accurately measure the tank level. The agitator was run at varying speeds and with varying grout viscosities to determine what parameters cause vortex formation and whether measurement accuracy is affected. (2) A dry feeds over addition. Engineering Calculating X-ESR-Z-00017 1 showed that an additional 300 pounds of dry premix added to a 300 gallon working volume would lower the water to premix ratio (W/P) from the nominal 0.60 to 0.53 based on a Salt Waste Processing Facility (SWPF) salt simulant. A grout with a W/P of 0.53 represents the upper bound of grout rheology that could be processed at the facility. A scaled amount of dry feeds will be added into the hopper to verify that this is a recoverable situation. (3) The necessity of baffles in the hopper. The preference of the facility is not to have baffles in the hopper; however, if the initial testing indicates inadequate agitation or difficulties with the radar measurement, baffles will be tested.

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Commercial submersible mixing pump for SRS tank waste removal - 15223

The Savannah River Site Tank Farms have 45 active underground waste tanks used to store and process nuclear waste materials. There are 4 different tank types, ranging in capacity from 2839 m{sup 3} to 4921m{sup 3} (750,000 to 1,300,000 gallons). Eighteen of the tanks are older style and do not meet all current federal standards for secondary containment. The older style tanks are the initial focus of waste removal efforts for tank closure and are referred to as closure tanks. Of the original 51 underground waste tanks, six of the original 24 older style tanks have completed waste removal and are filled with grout. The insoluble waste fraction that resides within most waste tanks at SRS requires vigorous agitation to suspend the solids within the waste liquid in order to transfer this material for eventual processing into glass filled canisters at the Defense Waste Processing Facility (DWPF). SRS suspends the solid waste by use of recirculating mixing pumps. Older style tanks generally have limited riser openings which will not support larger mixing pumps, since the riser access is typically 58.4 cm (23 inches) in diameter. Agitation for these tanks has been provided by four long shafted standard slurry pumps (SLP) powered by an above tank 112 kW (150 HP) electric motor. The pump shaft is lubricated and cooled in a pressurized water column that is sealed from the surrounding waste in the tank. Closure of four waste tanks has been accomplished utilizing long shafted pump technology combined with heel removal using multiple technologies. Newer style waste tanks at SRS have larger riser openings, allowing the processing of waste solids to be accomplished with four large diameter SLPs equipped with 224 kW (300 HP) motors. These tanks are used to process the waste from closure tanks for DWPF. In addition to the SLPs, a 224 kW (300 HP) submersible mixer pump (SMP) has also been developed and deployed within older style tanks. The SMPs are product cooled and product lubricated canned motor pumps designed to fit within available risers and have significant agitation capabilities to suspend waste solids. Waste removal and closure of two tanks has been accomplished with agitation provided by 3 SMPs installed within the tanks. In 2012, a team was assembled to investigate alternative solids removal technologies to support waste removal for closing tanks. The goal of the team was to find a more cost effective approach that could be used to replace the current mixing pump technology. This team was unable to identify an alternative technology outside of mixing pumps to support waste agitation and removal from SRS waste tanks. However, the team did identify a potentially lower cost mixing pump compared to the baseline SLPs and SMPs. Rather than using the traditional procurement using an engineering specification, the team proposed to seek commercially available submersible mixer pumps (CSMP) as alternatives to SLPs and SMPs. SLPs and SMPs have a high procurement cost and the actual cost of moving pumps between tanks has shown to be significantly higher than the original estimates that justified the reuse of SMPs and SLPs. The team recommended procurement of 'off-the-shelf' industry pumps which may be available for significant savings, but at an increased risk of failure and reduced operating life in the waste tank. The goal of the CSMP program is to obtain mixing pumps that could mix from bulk waste removal through tank closure and then be abandoned in place as part of tank closure. This paper will present the development, progress and relative advantages of the CSMP. (authors)

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Mixed-layered bismuth-oxygen-iodine materials for capture and waste disposal of radioactive iodine

Materials and methods of synthesizing mixed-layered bismuth oxy-iodine materials, which can be synthesized in the presence of aqueous radioactive iodine species found in caustic solutions (e.g. NaOH or KOH). This technology provides a one-step process for both iodine sequestration and storage from nuclear fuel cycles. It results in materials that will be durable for repository conditions much like those found in Waste Isolation Pilot Plant (WIPP) and estimated for Yucca Mountain (YMP). By controlled reactant concentrations, optimized compositions of these mixed-layered bismuth oxy-iodine inorganic materials are produced that have both a high iodine weight percentage and a low solubility in groundwater environments.

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